ML17191A301

Final Safety Analysis Report

ML17191A301
Person / Time
Site:	Dresden
Issue date:	07/10/2017
From:	Commonwealth Edison Co
To:	Office of Nuclear Reactor Regulation
References
Download: ML17191A301 (76)

Text

f FSAR INDEX

. ~.

- A- Section I *. - .~ ~->.

ACAD/CAM 6.8.3.3 Acceleratioa Response Spectrum Earthquake 12. 1. l :3 Access Control Access to the Site Access, Station Accident Analysis for 7 x 7 Fuel Accident Analysis for G2 8 x 8 Fuel 14.2.2.5 Accident Analysis for ENC 9 x 9 Fuel 14.2.2.6 Accident Analysis, Radwaste 9.2.5 Accident, Control Rod Drop Procedural 14. 2 .1. 3 Acoustic Monitors 4.5.2 Acronyms and Initialisms 1.1.2.1 Action taken due to Reportable 13.6.2.2 Action taken due to Safety Exceeded 13.6.2.1 Administrative Controls 13.6 Administrative 12 .1. 4. 5 Admission Valves 6.2.3.4 Airborne Effects the Refueling Pool 14.2.2.6 Air Cleanup Appendix 8 (8-28)

Air Ground Level Appendix A (2.1.1)

System 10.11

11. 2. 2 Ejector Off-Gas Monitoring 7.6.2.3 Monitoring, Reactor Bldg 7. 6. 2. 5 Airlock Doors 5.3.2.2 i

1043v

' FSAR INDEX

- A- Section Analysis and Acceptance Criteria Inst & Control 7.2.6.3 Analysis of Off-Site Electric Power Supply 8.2.1.4 Analysis Supporting ECCS Clad Melt Criteria 6.2.7.6 Analytical Methods 3.3.3 Analytical Stability Model 7.2.2.3 ANL Test Data on Clad Flailure 6.2.7:25-28 Approval of Changes 13.6 APRM 7.4 .2 Archifect - Engineer Organization Appendix E (2.3.1)

Area Radiation Monitoring System 7.6.3 9.1.2 9.5.3 As-Built Safety-Related Piping Analysis 12 .1. 2 .4 ASKE Class A Nuclear Vessels 4.1.0.1 Atmospheric Control System 6.8 Atmospheric Pressure, Fuel Loading 13.8.2.1 Atmospheric Weather/Wind Appendix G Authority to Terminate Power Production 13.6.1 Authorization of Changes 13.6 Automatic Depressurization System 6.2.6 Automatic Vacuum Relief 5.2.2.9 Auxiliary and Emergency Systems 10.1 13.7.3.42 Auxiliary Power supplies 1.2.4.3 Auxiliary Power System 8.2.1.3 Auxiliary Systems 1.2.4.4 Auxiliary Transformers 8.2.1.3 ii 1043v

FSAR INDEX

- A- Section Auxiliaries, Turbine Generator 13.7.3.43 Availability Analysis 6.2.7.4 Average Power Range Monitor (APRM) 7.4.5.2 iii 1043v

• FSAR INDEX

- B- Section Balance of Plant - Aux Systems 13.7.3.42 Bases for Design 12 .1.1. 3 Biological Shield 12.2.2.l Batteries, Station 8.2.3.2 Battery Tests and Inspection 8.3 Bio - Assay and Medical Exam Program 9. 5. 5. 7 Bodega Bay Tests 5.2.3.5a & b Boron 9.6.1.3.2 Blowoff Details, Rx Bldg. 5.3.2:1 Burnable Neutron Absorber 3.5.5 Burning in Drywell 6.8.1:12 Bypass Valves, Turbine 7.2.6.2

• 1057v i

FSAR INDEX

- c -

Section Cable Pans, Electric 8.2.2.3 Cask Pad 10 .1.2 CB & I 5.2.3:24 & 25 CECO and GE Startup Organization 13 .1. 2 Channel Hydrodynamic Conformance 7.2.3.2 7.2.4.2 Change Room Facilities 9.5.5.4

/

-characteristics After Reactor Slowdown 5.2.3.3 Charcoal Beds, Off-Gas 9.2.5 CHASTE 6.8.3.3.4 Chimney 12 .1. 2 .3 Chimney Effluent Monitoring 7.6.2.4 9.1 & 9.2.2.2 Circuit Breakers 8.2.2 Circulating Water 11. 2. 2 Cladding Integrity Safety Limit (Fuel) 3.2.2.3 Class I Structures & Equipment 12 .1. 2 Class II .Structures & Equipment 12 .1. 3 Classification of Nuclear Systems Appendix E (Exhibit 2. 7)

Cleanup Demineralizer System 13.7.3.22 Cleanup System 10.2 Cleanup System (Rx Water) 10.3 C02 Fire Protection System 10.7.2:1 & 2 Coefficiency of Reactivity 3.3.5.1 Cold Loop Startup - Transient Analysis 4.3.3:lla .& b

• Common Auxiliary Systems 1058v i

1.2.4.4

FSAR INDEX

- c -

Sectfon Conununication System "' 10.14 Computer, Process 7 .11 8.2.2.4 CONCEN 6.8.3.3.4 Conclusions on Site and Environs 2.4 Condensate Demineralizer System 7.8.2 13.7.3.13 Condensate - Feedwater System 11.1 11.3 Condensate - Feedwater Tests and Inspections 11.3 Condensate Makeup Piping 10.12.2:2 Conduct of Operations 13.1 thru 8 Conduct of Operations 13.1 Construction Tests 13.7.3

• Containment Containment Atmospheric Control System 1.2.1.3 5.2.3:7

7. 7. 2: 1 6.8 I

Containment Cooling System 6.2.4 Containment Design Basis Appendix 8 (B-26)

Containment Heat Removal Systems Appendix B (B-26)

Containment Isolation Valves 5. 2 .4.3; Appendix B ( B-2 7)

Containment Leakage Rate Testing Appendix B ( B-27)

Containment Penetrations 5.2.4.2 Containment Response to LOCA 5.2.3.2 Containment Shield 12.2.2.2

• • Containment Spray System 1058v ii 13.7.3.34 6.2.4.2.2

FSAR INDEX

- c - Section

• Containment Systems Containment Ventilation System i.2.2.4 5.l;*Appendix C 5.2.4.4 Containment Vs Hydrogen 6.8.1.3 Contractors 1.3 Control and Instrumentation 1.2.1.4 1.2.2.6 Control and Instrumentation, other Systems 7.10 Control Curtains 3.5.2.2 Control of Access to Radiation Zones 9.5.5.l Control Methods (Reactor) 3.5.2 Control Rods 3.5.2.1 Control Rod Block Function 7.3.2:1

• Control Rod Drive Control Rod Drive Housing Control Rod Drive Housing Supports 13.7.3.21 10.6.3 6.6 Control Rod Drive Housing Support Inspection & Testing 6.6.4 Control Rod .. i>ri ve Hydraulic System 10. 6 13.7.3.17 Control Rod Drive Mechanism 3.5.3.2 Control Rod Drive System 10.6.2:1 Control Rod Drop 14.2.1 Control Rod Drop Accident Procedural Safeguards 14.2.1.3 Control Rod Housing Support 6.1.2.4 Control Rod Hydraulic System 13.7:3.17 Control Rod Isometric 3.5.2:1

• Control Rod Movement 7.3.2 iii 1058v

FSAR INDEX

- c - Section

• Control Rod Sequence Control Rod Surveillance and Testing Control Rod Worth 14.5.2 3.5.4 3.3.4.4 Control Rod Velocity Limiter 6.1.2.3 6.5 Control Room 12 .1. 2. 2 12.2.2.4 12.2.3 14.2.5 Control Room Ventilation 12.2.2.5 Cooling Lake . 2.2.4.1 2.2.1:2 2.2.4:1 Core Cooling 14.2.3.9 Core Cooling System 6.2 Core Internals, Thermal Shock Efforts 3.6.3.3 Core Lattice Unit 3.4.2:2 Core Nuclear Dynamic Characteristic 3.3.5 Core Release, Non-Line Break Scenario 12.3.2.2 Core Spray Tests and Inspection 6.2.3.4 Core Spray System 6.2.3 13.7.3.32 6.2.3:6 8.2.3 Core Thermal and Hydraulic Performance 14.5.4 Crane, Reactor Building 10.1.2.2.2 Crib.House 2.2.6:2 12 .1. 3 ..3 Criteria & Bases for Design 12 .1.1. 3 CPR Histogram for 8 x 8 3.2.2:2

• 1058v ii i i

FSAR INDEX

- D- Section

• Data Analysis and Acceptance Criteria DC Systems Decay Ratio 7.2.6.3 13.7.3.2 & 8.2.3.2 7.2 Dernineralizer System 12.2.2.7 Description of Control Rods 3.5.2.1 3.5.3 Description of ECCS 6.2.2 Description of Fuel Assemblies 3.4.2 Description of Hain Stearn Description of Reactor Vessel Internals 3.6.2 Description of Safety Features 14.1 Design Basis Accidents 14.2 Design Basis Automatic Depressurization 6.2.6

• Design Basis Earthquake (Piping)

Design Basis of Core Spray Design Bases Dependent On Site Characteristics

12. 1. 2. 4 .,4 6.2.3 1.2.2.1 Design Basis of Fuel Mechanical Characteristics 3.4.1 Design Basis of Isolation Condenser 4.6.1 Design Basis of LPCI 6.2.4.1 Design Basis of Hain Stearn 4.4.1 Design Basis of Nuclear Characteristics, 3.3.1 Design Basis of Primary Containment System Design Basis o~ Reactivity Control Mechanical Characteristics 3.5.1 Design Basis of (Reactor) 3.2.1.1 3.2.1.3

• 1045v i

FSAR INDEX Section

- D-Design Basis of Reactor Bldg. 5.3.1 Design Basis of Reactor Recirculation System 4.3.1 Design Basis of Reactor Vessel Internals 3.6.1 Design Basis of Relief and Safety Valves 4.5.1 Design Bases for Shielding 1.2.2:1 Design Evaluation Containment System 5.2.3 Design Evaluation (Fuel) 3.4.3 Design Evaluation Main Stearn 4.4.3 Design Evaluation Reactor Coolant System 4.2.3

• Design Guide Limit Definition 7.2.4.1.

Design of Control Rods and Curtains 3.5.2.3 Design of Electrical Systems 8.2

• Design Report, Reactor Designed Safeguards Determination of Radiation Environment Appendix D

14. 2 .1. 2 12.3.3.0 Development of Technical Spec 3.2.4 Diesel - Generator System 8.2.3.1 8.3.1 13.7.3.39 Diesel Generator Tests and Inspection 8.3 Discharge to the River 9.3.3 Distances From Release Points 2.2.5:1 Distribution System, Station 8.2.2 Domestic Water Syste~ 13.7.3.8 Doppler Coefficient 3.3.5:1,2,3,4,5

• Dose, External Appendix A ( 2. 2 .1) ..

ii 1045v

FSAR INDEX

- D - Section

• Dose, Hydrogen Addition Dose to the Control Room, etal Dresden Lock and Dam

14. 2 .1.8

12. 3 .8.

2.2.6.1 Dresden Containment Certification Appendix C Dresden Units 2 & 3 Opera~ing Map 3.2.3.1 Dropout Velocities 6.5.3 Drywell 5.2.2.1 5.2.4.1 5.2.3.26 Drywell Pneumatlc System 10.8.2 Drywell and Suppression Chamber Inspection and Testing 5.2.4 Drywell Expansion Gap 5.2.3.6 Drywell Missile Protection 5.2.3.7

• Drywell Spray Drywell - Torus Leak Rate Measurement Drywell Ventilation 6 .2 .4 .2 ..

13.7.3.18 13.7.3.40

• 1045v iii

FSAR INDEX

- E- Section

• Earthquake Earthquake Analysis of Rx Vessel 5.2.3:8+9 12 .1.1: 2 Appendix D ECCS 1.2.2.5 6.2 6.2.7.5 Appendix B (B-23&25)

ECCS Clad Melt Criteria 6.2.7.6 ECCS*Flood Protection 6.2.8 ECCS Pipe Whip Criteria 6.2.7.7 ECCS Pump NPSH 6. 2. 7 .9 Economic Generation Control 7.3.3.l 7.3.6 Effect of KSIV Closure Time 14.2.3.8

• Effects of Postulated LOCA's EGC Operation El Centro Earthquake 1.2.5.2 7.3.3.2 7.3.6:2 12 .1.1: 2 Electrical Penetration Seals 5.2.2.4 5.3.2.3 Electric Power 1.2.1.5 Electric System 1. 2. 2 .10 8.1 Electroslag Weld Report, Rx Vessel Appendix F Elevated Release Point Discharge i4. 2 .1. 7 Emergency Core Cooling System 6.1.2.1 6.2.2 6.2.7.1 Appendix B (8-25)

• 1059v i

FSAR INDEX

- E - Section

• • Emergency Lighting Emergency Power Emergency Ventilation 10.13.2 8.2.3 13.7.3.4.1 Engineered Safey Features Appendix B (B-21&24)

Environs Radioactivity Monitoring 2.3 Equipment Description, Computer 7 .11. 3 Equipment Drain System 9.3.2.1 Equipment Separation 12 .1.4. 4 Equipment Supply - QA Appendix E (3.3)

Essential Service System 8. 2 .2*.4 Exclusi_on Area 2.2.1.6 Exfiltration 5.3.3.l Expansion Gap, Drywell

• 5.2.3.6 External Dose Appendix (2.2.1) ii 1059v

FSAR INDEX

- F Section Features of Plant Design 1.2.3:1 Feedwater Control System 7.8.3 Feedwater Flow, Reactor 7.5.2.4 Feedwater Nozzle Inner Bore 6.2.5.3.4 Feedwater Pumps 7.2.6.2 Feedwater Sparger Integrity 6.2.5.3.4 Feedwater System 11.1 11.3 14.2.3.5 Field Change Control Appendix E (3.4.3)

Fire Alarm Systems 10.14.3 Fire Extinguishers, Portable 10.7.2 Fire Protection System 10.7

• Fire Suppression Water System Fission Product Release from the Fuel 13.7.3.11 8.2.2.1 10.7.2 & 10.7.3 14.2.4.2 Fission Product Transport 14. 2 .1. 6 Flange Leak Detection, Reactor Vessel 7. 5 .2 .6 Floor Drain Surge Tank 9.3.2:5I Floor Drain System 9.3.2.2

'Flow Control Recirc System 7.3.3 Flow Factor, *Kf 3.2.2.9 Flow Monitors (Recirculation) 7.4.5.2.2 Flow Regulating Station (Circulating Water/Canal) 2.2.4 Fluid Pipe Penetration 5*. 2. 2. 5

• Flux Response to Rods 14.5.3 i

1060v

I~

FSAR INDEX

- F - Section

• Fractional Control Rod Density FSAR Controlled Copy Recipient Fuel and Waste Storage Systems 3.3.4:4 1.1.1.4 Appendix B (B-29}

Fuel Assembly Isometric 3.4.2:1 Fuel Cladding Integrity Safety Limit 3.2.2.3 3.2.4.2 Fuel Cycle 3.3.4.1 Fuel Damage Limits 3.2.1.2 3.4.3.4 14.2.2.5.1 14.2.2.6 Fuel Design Analysis 3.4.3.3 Fuel Handling 10.1 13 .1. 3. 2 13.7.3.20

• Fuel Handling and Storage Fuel Loading Fuel Mechanical Characteristics 1.2.1.8 1.2.2.8 13.8.2.1 3.4 Fuel Pool Cooling and Cleanup System 10.2 13.7.3.19 Fuel Pool Damage Protection 10 .1.4 Fuel Recovery Plant Appendix A (4.0}

Fuel Shipping Cask 10 .1. 2 .2 .2 10.1. 2 .3 Fuel Storage and Fuel Handling 10.1 Fuel Storage Criticality Appendix B (B-30}

Fuel Storage Pool (Spent} 10.1. 2. 2

-- Fuel Storage Vault 1060v ii 10.1. 2 .1

FSAR INDEX

- G - Section

• - Gadolinium Bearing Rods Gaseous Radioactive Wastes Gaseous Waste Effluents 3.5.5.5 9.1 1.2.4.1 9.2 GE Startup Organization 13 .1. 2 .1 General Arrangement Crib House 12 .1. 3 :8 General Arrangement, Rx Bldg. 12.1.2:1-4 General Arrangement, Turb. Bldg. 12.1.3:1 General Conclusions 1.4 General Description (Reactor) 3.3.2 General Electric Safety Analysis 14.3 General Electric Topical Reports 1.1.2.1 Generating Station Emergency Plan (GSEP) 13.4.1

• Generator Load Rejection Gee;> logy 11.2.3.2 7.7.1.2 2.2.3 Ground Level Radiation Dose- Appendix A (2.0)

Guide T~bes, CRD 6.5.2

• 1067v i

FSAR INDEX

- H- Section Halon System 10.7.2 Head Cooling System (Rx) 10.5 13.7.3.26 Health Physics 7.6.5 9.5.5 9.5.5.5 Health Physics Instrument Inspection and Testing 7.6.5.3 Heat Generation Rate 3.2.2.2 3.4.3.2 Heating Boiler 13.7.3.14 Heating, Ventilating, and A-C System 10.11 Heat up 13.8.2.2 High Density Spent Fuel Storage Rack 10 .1. 2: 2 High Neutron Flux 7.7.1.2

• High Primary Containment System Pressure High Radiation Sampling System (HRSS)

H,igh Reactor Pressure 7.7.1.2 9.6 7.7.1.2 Histogram of XN-3 Predictions 3.2.2:11 HPCI 6.2.5 13.7.3.33 6.2.5:1-5 8.2.3 HPCI Room Coolers 10.9.3 HPCI Tests and Inspection 6.2.5.4 HRSS 9.6 Hydraulic Control System (CRD) 3.5.3.3 Hydraulic (Reactor) Characteristics 3.2 Hydro Tests 13.7.3.16

• Hydrodynamic Stability 1046v i

7.2.2.2

FSAR INDEX

- H- Section

• Hydrogen Addition Hydrogen from Radiolysis Hydrogen from -H20 Reactions

14. 2 .1.8 6.8.1.2 6.8.1.1 Hydrogen in Containment Effects 6.8.1.3 Hydrology 2.2.4 Hypochlorite Chemical 10.9.2

• 1046v ii

FSAR INDEX Section

- I -

Identification, CRD 14. 2 .1.1 Identification of Contractor 1.3 IEEE 279 7.4.5 Impact Forces 14.2.3.7 Industrial Facility Near Station 2.2.2:2 In-Core Probe (TIP) 5.2.2.7 8.2.2.3 Inerting System 6.8.3.2 Initial Operating Personnel 13 .1.4 .1 Initialisms and Acronyms 1.1.2.1 Inservice Inspection 4.3.4.2 Inspection and Testing of Condensate and Feedwater 11.3 Ins~ection and Testing of Core Spray 6.2.3.4 Inspection and Testing of CRD Housing Support 6.6.4 Inspection and Testin_g of Diesel Generators and Batteries 8.3 Inspection and Testing of Drywell and Suppression Chamber 5.2.4 Inspection and Testing of Health Physics Instruments 7.6.5.3 I

Inspection and Testing of HPCI 6.2.5.4 Inspection and Testing of Isolation Condenser 4.5.4 Inspection and Testing of Low Pressure Coolant Injection 6.2.4.4 Inspection and Testing of Off gas and Ventilation 9.2.4 Inspection and Testing of Main Steam 4.4.4 Inspection and Testing of Reactor 3.6.4

• • 106lv i

Cr FSAR INDEX

- I - Section Inspection and Testing of Reactor Coolant ~ystern 4.2.4 4.3.4 Inspection and Testing of Reactor Vessel 4.2.4 Inspection and Testing of Recirculation System 4.3.4 Inspection and Testing of Safety and Relief Valves 4.4.4 Inspection and Testing of Secondary Containment 5.3.4 Inspection and Testing of Standby Coolant Supply 6.3.4 Inspection and Testing of Standby Liquid Controi System 6.7.4 Inspection and Testing of Stearn Flow Restrictors 6.4.4 Inspection and Testing of Turbine 11. 2. 4 Inspection, Weld, Visual 12.1.2.4.4.1 Institutional Facilities Near Station 2.2.2:3 Instrument and Service Air System 10.8 13.7.3.12 Instrumentation and Control 7.1 Instrumentation and Control-Containment 6.8.3.4 Integrated Plant Safety Assessment etal (IPSEP) 14.4.0 Integrated System Design Evaluation 6 ~ 2. 7 Inter-Plant Effects of Accidents 1.2.4.5 Interaction of Units 1,2, & 3 1. 2 .4 Interconnection, Electrical Network 8.2.1 Intermediate Range Monitor (!RM) 7.4.4 Introduction and Summary 1.1.

Iodine Activities 9.2.5 Iodine (I-131) Release Appendix A (3-4)

IRM 7.4 ii 106lv

I FSAR INDEX

- I - Section Isokinetic Sample 7.6.2.4.2 Isolation Condenser Inspection and Testing 4.6.4 Isolation Condenser Vent Monitor 7.6.2.9 Isolation Condenser - Piping Diagram 4.6.2:1 Isolation Valves 5.2.2.6 5.2.4.3 13.7.3.18 Appendix B (B-27)

Isotope N16 7.6.2 Isotopes in Liquid Waste Discharger 9.3.3 Investigative Function 13.6.2

• 106lv iii

FSAR INDEX

- J - Section

• Jet Pump Efficiency Jet Pump Isometric Jet Pump Operation 4.3.3.1 4.3.2:2 4.3.2.2 Jet Pump Stability 4.3.3.2

• 1047v i

FSAR INDEX

- K- Section

• 1048v i

FSAR INDEX

- L - Section

• Laboratory Radiation Measuring Inst Lake 7.6.5 2.2.4.1 2.2.1:2 Land Use 2.2.2.2 Leakage of Reactor Internals During Rec ire Line Break . 3.6.3.5 Leakage Rat~ Test, Rx Bldg 13.7.3.41 Lighting System 10.13 Limiting Safety System Settings 3.2.4.1 Liquid Radioactive Waste Discharge Monitorln~ 7.6.2.8 9.3 Liquid Waste Effluents 1.2.4.2 Liquid Waste Performance Analysis 9.3.3

\

Load Diagrams 12 .1. 2. 28

• Load Set Mechanism LOCA's 7.3.3.2.C 1.2.5.2 5.2.3:2 Loe.al Limits During Operations 3.2.2*

Local Power Range Monitor (LPRK) 7.4.5.1 Local Power Peaking 3.3.4.2 Lock and Dam 2.2.6.1 2.2.6:1 Loss-of-Control Room 14.2.5 Loss-of-Coolant Accident 14.2.4 Loss of EHC System Oil Pressure 11.2.3.2 7.7.1.2 Loss of Feedwater 11:3.3:2-3C Low Reactor Water Level 7. 7 .1: 2

• 1062v i

FSAR INDEX

- L - Section

• LPCI 6.2.4 13.7.3.33 6.2.4:1-6 8.2.3 LPCI Inspection and Testing 6.2.4.4 LPCI Room Coolers 10.9.3 LPRM 7.4.5:2-8 7.4

• *1062v ii

FSAR INDEX

- K- Section Kain Condenser Condensate 7.8.2 Kain Steam 4.4 14.2.3:1 Kain Steam Flow Restrictors 6.4 Kain Steam Isolation Valve 5.2.2:9 7.7.2:2 14.2.3:1 11.2.3:4-6 "L--.

Kain Steam Line Break Outside Drywell 14.2.3 Kain Steam Line Flow Restrictor 6.4.3:1

• Kain Steam Line Isolation Valve Closure 14.2.3.3 Kain *Steam Line Koni toring 7.6.2.2 Kain Steam Line Radiation Monitoring system 7.6.2:1 Kain steam Line Restrictors 6.1.2.2 6.4.2

• Kain Steam System Inspection and Testing Maintenance Department*

Makeup Water System 'j 4.4.4 13 .1. 3. 4 10.12 13 .. 7. 3. 8 MAPLHGR 7.4 Ka~t~r Flow Controller 7.3.3.2 Mathematical Model 12.1.2:5-7 Maximum Rate of Load Change 11.2.3.3 Maximum Recycle System 9.3.2:5J-M Maximum Rod Worth 3.3.4:6 KCPR 7.4 Mechanical Design Limits (Fuel) 3.4.3.1 Mechanical Vacuum Pump System 11.2 .2

• 1068v

/"

i

FSAR INDEX

- M- Section

• Medical Exam Program Metal-Water Reactions Meteorology

9. 5. 5. 7 5.2.3.4 2.2.5 Meteorological Factors Appendix A (2.1)

Midwest Fuel Recovery Plant Appendix A (4.0)

Minimum Shift Manning Requirements 13 .1.4. 2 Missile Protection Appendix B (B-25)

Mixture Impact Forces 14.2.3.7 Moderator Rod Worth 3.3.4:5 Moderator Temp. Coefficient of Reactivity 3.3.5:6 Moderator Void Coefficient of Reactivity 3.3.5:7 I-Monitoring Systems, Personnel 9.5.5.2 Motor - Generator Sets 7.3.3 Movement of Control Rods 7.3.2 MSIV 11.2.3.2 MSIV Closure Time 14.2.3.8

• 1068v ii

FSAR INDEX

' - N- Section

• N16 Isotope NOT Requirements 7.6.2 Appendix B (B-26)

Nearby Facilities - Potential Hazards 2.2.2.3 NEBS Instrumentation Systems 13.7.3.36 Negative Feedback 7.2.2.1 Network Interconnection 8.2.1 Neutron Flux Level 7.4.2 Neutron Monitoring Reliability 8.2.3.2.3 New Features 1. 2. 5 New Fuel Storage Vault 10 .1. 2 .1 Noble Gas Release Appendix A (3.3)

Normal Operation Characteristics 3.2.3 NPSH 4.3.2:3 NPSH for ECCS Pumps 6.2.7.9 NSS Supply, Material Appendix E (2.2.2)

NSS Periodic and On-Demand Programs, Computer 7.11.3.4 Nuclear Analysis Methods 3.5.5.4 Nuclear and Process Parameters 14.5 Nuclear Characteristics 3.3 Nuclear Instrumentation 7 .4 Nyquist Plot of Open-Loop Response 7. 2. 3: 7

• 1063v i

FSAR INDEX Section Off-Gas and Ventilation Inspection and Testing 9.2.4 Off-Gas Radiation Monitoring System 7.6.2:2 9.1 Off-Gas Treatment System 9.2.2:1 Off-Site Dose, Hydrogen Addition 14. 2 .1.8 Off-Site Electrical Power System 8.2.2.2 8.2.1.4 Off-Site Power and ECCS 6.2.7.5 Operability of the Units 1.2.5.3 On-Site Electrical Power System 8.2.2.1 On-Site Environs Radiation Monitoring System 9.5.4 Operating Basis Earthquake (Piping) 12 .1. 2. 4 Operating Basis (Reactor) 3.2.2.1

• Operating Group Operating Limit Heat Generation Rate Operating Limits (Reactor) 13 .1. 3 .1 3.4.3.2 3.2.1.3 Operating Procedures 13.3 Operational Description Recirc System 4.3.2.3 Operational Description of Recirculation Pumps 4.3.2.3 C & D Operational Design Guide and Conformance 7.2.4 Operational Training 13.2 Organization and Responsibility 13.1 Organization of Report 1.1. 2 Overall Quality Program . Appendix E (3.1) 138 KV System 8.2.1.3 13.7.3.3

• 1049v i

FSAR INDEX Section

• 115 Volt Systems 125 Volt DC Station Battery System

' 13.7.3.7 8.2.2:2

• 1049v ii

FSAR INDEX

- p - Section

• Partical Closure of Main Steam Line Isolation Valves Particulate Release (Sr-90) 7.7.1.2 Appendix A

( 3. 5) .

Peak Fuel Enthalphy 14.2.1:1-3 Pedestal, Reactor 12 .1.2. 5 Penetrations, Testing of Appendix B (8-27)

Performance Analysis (Rad Waste) 9.2.3 9.3.3 Performance Analysis (Shielding) 12.2.3 Performance Characteristic for Normal Operation 3.2.3 Performance Evaluation of Reactor Vessel, Internals 3.6.3 Performance Evaluation Recirc System 4.3.3 Performance *Predictions Recirc System 4.3.3.3 Peripheral Equipment, Computer 7.11.3.2 Personnel Monitoring Systems 9.5.5.2 13.4.2.2 Personnel Protection Equipment 9.5.5.3 13.4.2.3 Personnel Qualifications 13 .1. 4 Personnel Training 13.2.1:1 Physical Description Reactor Coolant System 4.3.2.1 Piping 12 .1. 2 .4 12 .1. 3 .4 Pipe Penetrations 5.2.2.5 5.2.4.2 5.3.2.3 Pipe Whip Criteria ECCS 6. 2. 7 .7 Plant Comparative Evaluation Appendix B i

1069v

FSAR INDEX

- p - Section

• Plant Description Plant Design Plant Effluents 1.2 1.2.3:1 Appendix B (B-31)

Plant Electrical Cabling 8.2.2.3 Plant Heating Boiler 13. 7 .. 3 .14 Plant Safety (SEP) 14.4.0 Plant Stability Analysis 7.2 Plot Plan 12.1.1:1 Plume Reflection Effects Appendix A (2.1.3)

Pool, Spent Fuel Storage 10 .1. 2 Population Data 2.2.2.1 2.2.2:1

• Portable Fire Extinguishers Portable Instrumentation Post-Accident Radiation Levels 10.7.2 9.5.5.6 12.3.1-1 Potential Hazards Due To Nearby Facilities 2.2.2.3 Power Flow Map 3.2.3:3 Power Range Instruments 7.4.5 Power Transient Analysis 14. 2 .1.4 Pre-Operational Training 13.2.1 Pre-Operational Test Program 13.7 Precautionary Planning 13.4 Pressure Forces During Blowdown (Reactor) 3.6.3.2 Pressure, Reactor Vessel 7.5.2.2 Pressure Regulator and Turbine-Generator Controls 7.8.1 ii 1069v-

FSAR INDEX

- p - Section

• Pressure Suppression Chamber Primary Containment Isolation Surveillance and Testing Primary Containment Isolation System 5.2.2.3 7.7.2.4 7.7.2 Primary Containment Sizing 5.2.3.1 Primary Containment System 5.2 Primary Piping 4.3.4.l Primary System Expansion 13.7.3.29 Primary System Hydro 13.7.3.16 Principal Design Criteria 1. 2 .1 Procedural Safeguards 14. 2 .1. 3 14.2.2.3 Procedure Designations and Categories 13.3.0:1 Process and Area Monitoring 9.1.2

• Process and Instr~mentation Process Computer

\

System Equip Chart 1.1.2:1 7 .11 8.2.2.4 Process Liquid Monitoring 7.6.2.7

.Process Radiation Monitoring 7.6.2 Property Plat 1.2.2:1 Protection E~uipment, P~rsonnel 9.5.5.3 Protection Systems 7. 7 Pump Back System 10.8.2 Purge, Vent, and Inerting System 6.8.3.2

• 1069v iii

FSAR INDEX

- Q- Section

• Quality Assurance Records Quality Control Reports Appendix E (3. 7)

Appendix E

• lOSOv i

FSAR INDEX

- R-Section Racks, High Density Spent Fuel Storage 10.1. 2 Radiation Control Standards 13.4.2 Radiation Dose (Fuel Pool) 10.1. 2. 2. 2 Radiation Levels, Post-Accident 12.3 Radiation Monitoring Systems / 1.2.2.7 2.3 7.6 7.6.4 Radiation Protection Procedures 1.2.2.11 Radiation Protection 9.5 Radiation (High) Sampling System 9.6 Radiation Shielding (HRSS) 9.6.3.0 Radiation Zones 9.5.5.1 Radioactive Waste Control 1. 2. 2 .12

• Radioactive Waste Disposal Radiological Effects 9.1 1.2.1.6 13.7.3.35

14. 2 .1. 5 14.2.3.10 14.2.4.2 Radiolo~ical Factors Appendix A (2.2)

Radiolysis 6.8.1.2 Radwaste Air Sparging System 10.8.2 Radwaste Building 12 .1. 3. 2 Radwaste Process Systems Radwast~ Ventilation 13.7.3.44 Ramp Rate 7.3.6.3 Rate of Response (CRD) 3.5.3.1

• 1064v i

FSAR INDEX

- R - Section RBCCW (Reactor Building Closed Cooling Water) 7.6.2.7 10.10 13.7.3.15 Reactivity Control 3.3.4.3 3.3.5.1 3.5 Reactivity Insertion Accidents 1.2.5.1 Reactor* Slowdown 5.2.3.3 Reactor Building 5.3 5.3.2.1 12 .1. 2 .1 Reactor Building Air Monitoring 7.6.2.5 Reactor Building Closed Cooling Water System 7.6.2.7 10.10 13.7.3.15 Reactor Building Crane 10 .1.2. 2 .2

• Reactor Building Leakage Rate Reactor Bldg Ventilation Reactor Building Ventilation Exhaust 5.3.4.1 13.7.3.41 13.7.3.44 9.2.2.1 Reactor Building Ventilation Isolation Valves 5.3.2.4 Reactor Building Ventilation Stack Monitoring

• 7.6.2.6 9.1 Reactor Coolant System 4.1 4.1.0:1 Reactor Coolant Slowdown 14.2.3.6 Reactor Coolant Pressure Boundary Appendix B (8-18&26)

Reactor Control Systems 7.3 Reactor Core* 1.2.1.1 Reactor Core and Channel Hydrodynamic Stability 7.2.2.2

• 1064v ii

7. 2. 3. 3

FSAR INDEX

- R- Section Reactor Core Conformance 7.2.4.3 Reactor Core Cooling System 1.2.1.2 Reactor Core Shutdown 14.2.3.4 Reactor Design Basis 3.2.1.1 Reactor Operating Limits 3.2.1.3 Reactor Pedestal 12 .1. 2. 5 Reactor Pressure Control 7.3.5 Reactor Pressure Vessel Design Appendix D Reactor Protection System 7. 7 .1 13.7.3.37 Reactor Protection System Surveillance and Testing 7.7.1.4 Reactor Recirculation System 13.7.3.31 Reactor Relief Valves 4.5.2 Reactor Shutdown Cooling System 10.4 Reactor Systems 1.2.2.3 3.1 Reactor Vessel 4.2 4.2.1:1 Reactor Vessel Components 13.7.3.27 Reac~or Vessel Designed Cycles 4.2.1:1 Reactor Vessel Ele~troslog Weld Report Appendix F Reactor Vessel Head Cooling System 10.5 13.7.3.26 7.6.2.7 Reactor Vessel Instrumentation Surveillance and Testing 7.5.4 Reactor Vessel Isometric 4.3.2:1 Reactor Vessel Hydro 13.7.3.16 Reactor Vessel Instrumentation 7.5 13.7.3.28 9'

iii 1064v

FSAR INDEX

- R- Section Reactor Vessel Internals 3.6 Reactor Vessel Lateral Supports 4.2.2:1 Reactor Vessel Nozzle Safe Ends 4.2.2.1 Reactor Vessel Inspection and Te~ting 4.2.2 Reactor Vessel Supporting Structure and Stabilizers 12 .1. 2. 5 Reactor Water Cleanup Piping Diagram 10.3.1:1 10.3.2 Reactivity Control Appendix B (B-15)

Recipient, FSAR Controlled Copy 1.1.1.4 Recirculation Flow Monitors 7.4.5.2.2 Recirculation Line Break 3.6.3.5 Recirculation Pumps Operational Description 4.3.2.3.C & D Recirculation Speed Control Network 7.3.3:1 Recirculation System 4.3 13.7.3.31 Recirculation System Analysis 4.3.3.4 Recirculation System Inspection and Testing 4.3.4 Records 13.5 Appendix E (3.7.1)

Recreational Facility Near Station 2.2.2:3 Refueling 10 .1.2 .3 Refueling Accident 14.2.2 Refueling Accident Procedural Safeguards 14.2.2.3 Refueling Pool Airborne Effects 14.2.2.6 Regional and Site Meteorology 2.2.S Relative Bundle Power Histogram 3.2.2:1 & 3 i i ii 1064v

FSAR INDEX

- R- Section

• Release of Activity to Environment (Liquid)

Relief and Safety Valves 9.3.3 Appendix B (B-31) 4.5 13.7.3.30 Reliability of Protection Systems Appendix B

( B-12 )"

Reportable Occurrence 13.6.2.2 Resumes of Startup Personnel Appendix H Review and Investigative Function 13.6.2 Ring Header 5.2.3:18-23 Rod Block Monitor (RBM) 7.4.S.3 7.4.S.4 Rod Drop Accident Analysis 12 .1.4. 6 14.2.1:4 Rod Movement Tests 7.2.6.2

• Rod Worth Mini~izer 7.9 13.7.3.38

\

• 1064v iii ii

FSAR INDEX

- T - Section

• T-Quencher Technical Spec. Development Technical Staff 4.5.2 3.2.4 13 .1.3. 3 Temperature, Reactor Vessel 7.5.2.1 Test Schedule, Pre-operational 13.7.2 Testable Check-Isoiation Valves 6.2.3.4 Testing and Surveillance (Reactor) 3.4.4 Thermal (Reactor) Characteristics 3.2 Thermal Shock Effect*s on Core Internals 3.6.3.3 Thermal Shock Effects on Reactor Vessel Components 3.6.3.4 Thermal Sleeves, Feedwater Nozzle 6.2.5.3.4 TIP 7.4.2 Topical Report (CECo) 13.2.2 Topical Report (GE) 1.1.2.1 Tornadoes 2.2.5.l Torus 5.2.2.3 5.2.3:17 Torus Seismic Analysis 5.2~3:2 Torus Water Contamination 6.2.7.8 Total System Conformance 7.2.3.4 7.2.4.4 Transient Operating Conditio~s 3.2.4.3 Traversing Incore Probe (TIP) 5.2.2.7

7. 4. 5. 5 Trend Records 7.11.3.3 Turbine 11. 2. 2 Turbine Building 12 .1. 3 .1 i

1052v r

FSAR INDEX

- T - Section Turbine Building Cooling Water System 13.7.3.10 10.9.2 Turbine Building Ventilation 13.7.3.44 Turbine Bypass System 11. 2. 2 Turbine Condenser 11. 2. 2 Turbine Generator 11.2 13.7.3.43 Turbine Generator Controls 7.8.1 Turbine Generator System 11. 2. 2 Turbine Plant Control Systems 7.8 Turbine Steam Handling Equipment 12.2.2.6 Turbine Stop and Bypass Valves 11. 2 .4 Turbine Stop Valve Closure 7.7.1.2 Turbine System 1.2.2.9

• Turbine Tests and Inspection 11. 2 .4 Turbine Trip With Flux Scram 4.5.3:2a&b

11. 2. 3 Turbine Trip Without Bypass 3.2.2:10 Turnkey Projects Operation Appendix E (2.2-1)

Typical Core Lattice Unit 3.4.2:2 345 KV System 8.2.1.2 13.7.3.4 220 Volt and 115 Volt Ac Systems 13.7.3.7 250 Volt DC Station Battery System 8.2.2:1 ii l052v

FSAR INDEX

- s - Section

• Standby Lighting Standby Liquid Control System

. 10.13 .3 6.7 7.3.4 13.7.3.25 Standby Liquid Control System Inspection and Testing 6.7.4 Startup and Power Test Program 13.8 Startup Program, Preoperational 13.7.1 Startup Tests Inst and Control 7.2.6.2 Station Access 13.4.3 Station Arrangements - 1.2.2.2 2.2.1:1 station Batteries 8.2.3.2 8.3.2 Station Computer Power Supply 8.2.2.4 Station Distribution System 8.2.2

• Station Fire Protection System Station Generated Procedures Station Grounding-Construction Tests 10.7 & 13.7.3.11 13.3 13.7.3.1 Station Instrument and Service Air System 10.8 Station Organization/Management 13 .1. 3 Station Procedure Designations and Categorie~ 13.3.0:1 Steady State 3.3.4 Steam Flow 7.5.2.5 Steam Flow Restrictors 6.4 Steam Handling Equipment, Turbine 12.2.2.6 .

Steam Jet Air Ejectors 11. 2. 2 Stock System 9.4.2.1 Structures anq Equipment 12 .1.1.1 iii 105lv

FSAR INDEX

- s - Section

• Structural Design and Shielding Stock Rod Margin Summary Evaluation of Safety

• 12.l 3.3.4:3 1.2.2.13 Summary of Off-Site Doses from Accidents 1.2.2:2 Summary of Pre-operational Test Content & Sequence 13.7.3 Summary of Technical Data 1.2.3 Supplementary Control 3.5.5 Suppression Chamber and Drywell Inspection and Testing 5.2.4 Surveillance and Testing of Control Rods 3.5.4 Surveillance and Testing of Nuclear Instruments 7.4.5.6 Surveillance and Testing of Primary Containment Isolation 7.7.2.4 Surveillance and Testing of Reactor 3.4.4 3.5.4 Surveillance and Testing of Reactor Protection System 7.7.1.4 Surveillance and Testing of Reactor Vessel Instrumentation 7.5.9 System Performance Transients 6.2.7.2

\

• 1051v iiii

FSAR INDEX

)*

- T - Section

• Technical Spec. Development Technical staff Temperature, Reactor Vessel 3.2.4 13 .1. 3. 3 7.5.2.1 Test Schedule, Pre-operational 13.7.2 Testing and Surveillance (Reactor) 3 .4 .4.

Thermal (Reactor) Characteristics 3.2 Thermal Shock Effects on Core Internals 3.6.3.3 Thermal Shock Effects on Reactor Vessel Components 3.6.3.4 Thermal Sleeves, Feedwater Nozzle 6.2.5.3.4 TIP 7.4.2 Topical Report (CECo) 13.2.2 Topical Report (GE) 1.1.2.1

• Tornadoes Torus Torus Seismic Analysis 2.2.5.1 5.2.2.3 5.2.3:17 5.2.3:2 Torus Water Contamination 6.2.7.8 Total System Conformance 7.2.3.4 7.2.4.4 Transient Operating Conditions 3.2.4.3 Traversing lncore Probe (TIP) 5.2.2.7

7. 4. 5. 5 Trend Records 7.11.3.3 Turbine 11.2 .2 Turbine Building . 12. 1. 3 .'1 Turbine Building Cooling Water System 13.7.3.10 10.9.2

• 1052v i

FSAR INDEX

- T - Section Turbine Building Ventilation 13.7.3.44 Turbine Bypass System 11. 2. 2 Turbine Condenser 11. 2. 2 Turbine Generator 11. 2 13.7.3.43 Turbine Generator Controls 7.8.1 Turbine Generator System 11. 2. 2 Turbine Plant Control Systems 7.8 Turbine Steam Handling Equipment 12.2.2.6 Turbine Stop and Bypass Valves 11. 2 .4 Turbine Stop Valve Closure 7.7.1.2 Turbine System 1.2.2.9 Turbine Tests and Inspection 11. 2 .4

• Turbine Trip With Flux Scram Turbine Trip Without Bypass 4.5.3:2a&b

11. 2. 3 3.2.2:10 Turnkey Projects Operation Appendix E (2.2-1)

Typical Core Lattice Unit 3.4.2:2 345 KV System 8.2.1.2 13.7.3.4 220 Volt and 115 Volt Ac Systems 13.7.3.7 250 Volt DC Station Battery System 8.2.2:1

• 1052v ii

FSAR INDEX

- u- Section Ultimate Performance Limit Criteria 7.2.3 Ultrasonic Resin Cleaners 9.3.2.4 Unit Auxiliary Power Supplies 1.2.4.3 Unit Control and Instrumentation 1.2.2.6 Unit-1 Spent Fuel 10.1. 2. 2 .1 Updated FSAR 1.1.1.3 1.1.1.4 i

1053v

FSAR INDEX

- v - Section

• Vacuum* Pump System Vacuum Relief Velocity Limiter, CRD

11. 2. 2 5.2.2.9 6.2.5 Vent Pipes 5.2.2.2 Vent, Purge, and Inerting Systems 6.8.3.2 Venting and Cooling System 5.2.2.8 Ventilating 10.11 Ventilation and Off-Gas Inspection and Testing 9.2.4 Ventilation, Control Room 12.2.2.5 Ventilation, Drywell 13.7.3.40 Ventilation, Emergency 13.7.3.41 Ventilation, Reactor, Radwaste, and Turbine Bldgs 13.7.3.44 Ventilation Stack Monitoring, Reactor Bldg 7.6.2.6 9.2.2.1 Ventilation System Containment 5.2.4.4 Venturis, Hain Steam Line 6.4.2 Vessel Components, Reactor 13.7.3.27 Vessel Head Cooling System 10.5 Vessel Instrumentation 13.7.3.28 Vibration of Components (Rx Internals) 3.6.3.1 Visual Weld Inspection 12.1.2.4.4.1.3 Vulkene Insulation 8.2.2.3 i

1054v

FSAR INDEX

- w- Section Waste Concentrator System 9.3.2.3 Water Level, Reactor Vessel 7.5.2.3 Water System (Clased Cooling) 10.10 Water System (Service) 10.9 Weather, Wind Appendix G Weld Inspection, Visual 12.1.2.4.4.1.3 Well Water System 10.12,2:1 Wind Appendix G WINDOW 6.8.3.3.4 i

1065v

FSAR INDEX

- x-Section Xenon Equilibrium 6.7.1 Xeno.n Stability 3.3.5.2 7.2.4.S X-Area Coolers 10.9.2 10.9.3

• lOSSv i

FSAR INDEX

- y - Section

• 1066v i

FSAR INDEX

- z- . Section

• -6. 8 .1.1

• 1056v i

v TABLE OF CONTENTS

- DRESDEN UNITS 2 & 3 UPDATED FINAL SAFETY ANALYSIS REPORT SECTION 1 INTRODUCTION AND

SUMMARY

2 SITE 3 REACTOR CORE AND INTERNALS 4 REACTOR COOLANT SYSTEM 5 CONTAINMENT SYSTEMS 6 ENGINEERED SAFEGUARDS 7 CONTROL AND INSTRUMENTATION 8 ELECTRICAL SYSTEM 9 RADWASTE SYSTEM 10 REACTOR AUXILIARIES 11 TURBINE AND CONDENSATE SYSTEMS 12 STRUCTURES AND SHIELDING 13 CONDUCT OF OPERATION

e. 14 SAFETY ANALYSIS APPENDIX A CHIMNEY RELEASE RATE CALCULATION B PLANT COMPARATIVE EVALUATION WITH DESIGN CRITERIA c CONTAINMENT CERTIFICATIONS D UNIT 2 REACTOR PRESSURE VESSEL DESIGN E QUALITY CONTROL F REACTOR VESSEL ELECTROSLAG WELD REPORT G METEOROLOGICAL DATA H RESUMES FOR STARTUP PERSONNEL

-e

Rev. 4 June 1986 1i TABLE OF CONTENTS SECTION 1 -- INTRODUCTION*AND

SUMMARY

1.1 PURPOSE AND ORGANIZATION OF REPORT l.Ll-1 1.1.1

• PURPOSE OF REPORT 1.1.1:-1 1.1.1.1 Introduction 1.1.1-1 1.1.

1.2 Purpose and Scope

of Safety Analysis Report 1.1.1-1 1.1.1.3 Updating of Original FSAR 1.1.1-2 1.1.1.4 FSAR Controlled Copy Recipient 1.1.1-2 1.1.2 ORGANIZATION OF REPORT 1.1. 2-1 1.1.2.1 General Format 1.1. 2-1 1.1.2.2 Revisions 1.1.2-1 1.2 PLANT DESCRIPTION 1. 2 .1-1 1.2 .1 PRINCIPAL DESIGN CRITERIA 1.2.1-1 1.2.1.1 Reactor Core 1.2.1-1 1.2.1.2 Reactor Core Cooling Systems 1. 2 .1-2 1.2.1.3 Containment 1. 2 .1-2 1.2.1.4 Control and Instrumentation 1. 2 .1-3 1.2.1.5 Electrical Power 1. 2 .1-3 1.2.1.6 Radioactive Waste Disposal 1. 2 .1-3 1.2.1.7 Shielding and Access Control 1. 2 .1-3 1.2.1.8 Fuel Handling and Storage 1. 2 .1-4 1.2 .2

SUMMARY

DESIGN DESCRIPTION AND SAFETY ANALYSIS 1.2.2-1 1.2.2.1 Design Bases Dependent On Site Characteristics 1.2.2-1 1.2.2.2 Station Arrangements 1. 2. 2-3 1.2.2.3 Reactor Systems 1.2.2-3 1.2.2.4 Containment Systems 1. 2. 2-4 1.2.2.5 Shutdown Cooling System and ECCS 1. 2. 2.,... 7 1.2.2.6 Unit.Control and Instrumentation 1. 2. 2-8 1.2.2.7 Radiation Monitoring Systems 1.2.2-9 1.2.2.8 Fuel Handling and Storage 1. 2. 2-9 1.2.2.9 Turbine System 1. 2 .2-10

1. 2. 2 .10 Electrical System 1. 2. 2-10 1.2.2.11 Shielding, Access Control, and Radiation Protection Procedures 1. 2. 2-10

1. 2. 2 .12 Radioactive Waste Control 1.2.2-11

1. 2. 2 .13 Summary Evaluation of Safety 1.2.2-11 1.2 .3

SUMMARY

OF TECHNICAL DATA 1. 2. 3-1

1. 2 .4 INTERACTION OF UNITS 1, 2, & 3 1.2.4-1 1.2.4.1 Gaseous Waste Effluents 1.2.4-1 1.2.4.2 Liquid Waste Effluents 1.2.4-1 1.2.4.3 Unit' Auxiliary Power Supplies 1. 2 .4-2 1.2.4.4 Common Auxiliary Systems 1. 2 .4-2 e 1.2.4.5 Inter-Plant Effects of Accidents 1. 2. 4-4 0013f OOOlf

1ii TABLE OF CONTENTS (Contd.)

SECTION 1 -- INTRODUCTION AND

SUMMARY

1.2.5 NEW FEATURES 1.2.5-1 1.2.5.1 Features l~hich Reduce the Probability and Magnitude of Potential Reactivity Insertion Accidents 1. 2. 5-1 1.2.5.2 Features Which Mitigate Effects of Postulated LOCA 1 s 1.2.5-1 1.2.5.3 Features Which Improve Operability of the Units 1.2.5-2 1.3 IDENTIFICATION OF CONTRACTORS 1. 3. 0-1 1.4 GENERAL CONCLUSIONS 1.4 .0-1

Rev. 2 June 1984 liv

• 1.1.2:1 1.1.2:2 LIST OF TABLES -- SECTION 1~ INTRODUCTION General Electric Company Topical Reports Acronyms and Initialisms 1.2.2:1 Design Bases For Shielding 1.2.2:2 - S1.DT1mary of Maximum Off-site Doses From Postulated Accidents 1.2.3:1 Principal Features of Plant Design

Rev. 1 June 1983

• . e .*LIST OF TABLES -~ SECTION 1~ INTRODUCTION liv LL2:1 1.2~2:1 General Electric Company Jbpical Reports Design Bases For Shfelding

• I 1.2.2:2 .* Summary of Maximum Off-site Doses From Postulated AcCidents 1.2.3:1 Principal Features of Pl~nt Design*

.. ~ .

\

t * * * *

. ~'. ... '* *'.

'* .')

~ r .:

. I

. *. .. . . .. . . ~ . *.. ....  :- ~*-* -~ ......... - .

l

,olJ **- ,..... ..w.--.~. **-*------** *--- ---

Rev. 4 June 1986 1.1.1-2

• 1.1.1.3 Updating of original FSAR This documertt is the Updated Final Safety Analysis Report {UFSAR), a report separate and distinct from the original Final Safety Analysis Report.

The original FSAR and the associated docket files {50-237 and 50-249) are the basis for the licensing of the plant. In the event that a discrepancy exists between the original FSAR and the UFSAR, the original FSAR will be the final authority. The Technical Specifications may reference the UFSAR.

The UFSAR is revised annually as required in 10 CFR 50.7le. The UFSAR is designe*d to serve as a reference document, reflecting the current configuration of the plant, including information and analyses required by and submitted to the NRC since submission of the original FSAR, and containing the information in a contiguous format.

1.1.1.4 FSAR Controlled Copy Recipient

Subject:

FSAR Update Dresden Station has reviewed the FSAR for rev1s1ons, corrections,

• and material information additions. The changes contained herein will become Revision 4 {June, 1986) to the FSAR .

The changes are in compliance with the 10 CFR 50.71{e) requirement to identify changes and which references the requirements defined in 10 CFR 50.59. The 50.59 report refers to changes in the facility as described in the FSAR, changes in procedures described in' the FSAR, and tests or experiments not described in the {original) FSAR.

All changes which have been implemented were previously reviewed to the 50.59 criteria and in our opinion do not constitute any additional unreviewed safety questions.

Dated n Manager Dresden Nuclear Power Station 0013f OOOlf

Rev. 3 June 1985 TABLE 1. 1. 2: 2

• APR APRM ASME ACRONYMS AND INITIALISMS Automatic Pressure Relief average power range monitor American Society of Mechanical Engineers BTP Branch Technical Position BWR boiling-water reactor CE Co Commonwealth Edison Company CFR Code of Federal Regulations CSE Containment Systems Experiments CST condensate storage tank CVTR Carolina Virginia Tube Reactor DBE design-basis event DER design electrical rating DG diesel generator ECCS emergency core cooling system EHC electrohydraulic control EI&C electrical instrumentation and control FSAR Final Safety Analysis Report FTOL full-term operating license FWCI feedwater coolant injection GDC General Design Criterion(a)

GE General Electric Company gpm gallons per minute

• HEPB hp HPCI IE IEEE IP SAR

~igh energy pipe break horsepower high-pressure coolant *injection Off ice of Inspection and Enforcement Instit~te of Electrical and Electronics Engineers Integrated Plant Safety Assessment Report IREP Integrated Reliability Evaluation Program IRK intermediate range monitor LCO limiting condition for operation LER licensee event report LOCA loss-of-coolant accident LPCI low-pressure coolant injection LPRM low power range monitor LWR light-water reactor MCC motor control center.

MCPR minimum critical power ratio MDC maximum dependable capacity MOV motor-operated valve mph miles per hour MSIV main steam isolation valve

• MSL mean sea level MWe megawatt-electric MWt megawatt-thermal NRC U.S. Nuclear Regulatory Commission ORNL Oak Ridge National Laboratory PMF probable maximum flood PMP probable maximum precipitation POL provisional.operating license 0013f OOOlf

Rev. 3 June 1985 TABLE 1.1.2:2 .(Cont'd)

• PRA psi psig PWR probabilistic risk assessment pounds per square inch pounds per square inch gage pressurized-water reactor RBCCW reactor building closed cooling water RCPB reactor coolant pressure boundary RPS reactor protection system RSCS reactor shutdown cooling system RWCU reactor water cleanup SALP Systematic Appraisal of Licensee Performance SAR safety analysis report SBGTS standby gas treatment system SEP Systematic Evaluation Program SER - safety evaluation report SOAD Station Operational Analysis Department SRP Standard Review Plan STS Standard Technical Specification

• • sws service water system TMI Three Mile Island UHS ultimate heat sink USI unresolved safety issue

• 0013f OOOlf

1.2.2-4 The core is assembled in modules of four fuel assemblies set in the interstices of a crucifonn control rod. This modular core fonn, common to all General Electric boiling water reactors, permits substantial increase in thennal power with a small increase in core diameter and at the same time preserves the reactivity control characteristics demonstrated in the

• operation of Dresden Unit 1 and other General Electric power reactors.

The reactor pressure vessel contains the reactor core and structure, steam separators and dryers, jet pumps, control rod guide tubes, and feed-water, emergency core cooling system (ECCS), and standby liquid control spargers and other components as shown in Figure 3.6.2:1. The inside diameter of the vessel is approximately 21 feet and the inside height between heads is approximately 68 feet. The main connections to the reactor vessel include the steam lines, jet pump lines, feedwater lines, and control rod drive thimbles. Other connections are provided for the isolation condenser system, standby liquid control system, ECCS, and in~trumentation syste~s.

The fuel for the reactor core consists of uranium dioxide pellets contained in sealed Zircaloy-2 tubes.

These fuel rods are assembled into square arrays in individual assem-blies. The original assemblies were of a 7x7 configuration; later designs introduced in subsequent fuel cycles were of an 8x8 configuration. The fuel enrichment is varied from rod to rod within an assembly to achieve desired neutron flux characteristics. Some water rods may be included, and gadolinium is used in some rods as a burnable poison, in the fonn of Gd 2 03 -U0 2

• Each fuel assembly is surrounded by a Zircaloy-4 flow channel.

Water serves as both the moderator and coolant for the core.

The control rods consist of assemblies of 3/16-inch diameter, sealed, stainless steel tubes filled with compacted boron carbide powder and held in a crucifonn array by a stainless steel sheath of 1/16 inch wall thickness fitted with castings at each end. The design of such control rods is almost identical with those which have been used successfully in Unit 1 for more than six years except that control rods of current design are longer *due to the use of longer fuel assemblies. The control rods are of the bottom entry type and are moved vertically within the core by individual, hydraulically operated, locking piston type control rod drives .

The control rod drive hydraulic system is designed to allow control rod withdrawal or insertion at a limited rate, one rod at a time , for power level control and flux shaping during reactor operation. Stored energy available fran gas charged accumulators and from reactor pressure provides hydraulic power for rapid simultaneous insertion of all control rods for reactor shutdown. Each drive has its own separate control and scram devices.

The systems for reactivity control are of the same design as those used in the Oyster Creek and Nine Mile Point Plants, including two features which provide improved plant safeguards.

1.2.3-1 1.2 .3

SUMMARY

OF TECHNICAL DATA Design features and data appropriate to achieve a reactor thermal output of 2527MW are summarized in Table 1.2.3:1.

TABLE 1.2.3:1 PRINCIPAL FEATURES OF PLANT DESIGN Site Location Dresden Site, County of Grundy, State of Illinois Size of Site 953 Acres plus 1275 acre cooling lake Site and Plant Ownership Commonwealth Edison Company Plant Net Electrical Output 809 MW Gross Electrical Output 850 mi Net Heat Rate 10,648 Btu/kw-hr Feedwater Temperature 340.1 F Thermal and Hydraulic Design Design Thennal Output 2527 M~*Jt

• Reactor Pressure (dome) 1020 psia 6 Steam Fl ow Rate 9.765 x 10 lb/hr Recirculation Flow Rate 98 x 10 6 lb/hr Fraction of Power Appear- 0.965 ing as Heat Flux GE GE GE 7x7 8x8 8x8R/P8x8R Power Density 41.08 kw~l i ter 41.09 40.74 Heat Transfer Surface Area/ 86.52 ft 97.6 94.9 Assembly 2 Average Heat Flux 131,200 Btu/(hr-ft 2 ) 117 ,100 120,400 Maximum Heat Flux 405,000 Btu/(hr-ft ) 354,400 362,000 Maximum U0 2 Temperature 3470°F Average Volumetric Fuel Temp. 1050°F Core Subcool i ng 22.4 Btu/lb Core Average Void Fraction, 0.299 Active Coolant Core Average Exit Quality 0.101 Minimum Critical Power Ratio 1.06 1.06 1.07 Safety Limit

1.2.3-2 TABLE 1.2.3:1 (Contd.)

PRINCIPAL FEATURES OF PLANT DESIGN Nuclear Design Enrichment No. of rods Wt % U-235 per assembly Initial Fuel Enrichment: 2.44 30

( 7x7 assembly) 1.69 16 1.20 3 Typical Reload Fuel Enrichment: 3.8 14 (8DRB265H 8x8 assembly) 3.0 27 2.4 2 2.0 14

1. 7 4 1.3 1 water rods 2 GE GE GE 7x7 8x8 8x8R Water/U0 2 Volume Ratio 2.41 2.60 2.76 Core Average Neutron Flux Thenna 1 3.50 x 10 13 13 n/cm 22-sec 1 Mev 3.67 x 10 n/cm -sec Burnup target (average assembly) 28 ,ooo MvJD/ton Power Coefficient for xenon stability More negative than

-.Ol(dK/K)/(dP/P)

Design Operating Heat flux peaking factors:

Relative Assembly 1.47 1.47 Axial 1.57 1.57 Local 1.30 1. 30 Overpower 1.20 Gross 3.60 3.00

. Reactivity Control:

Cold shutdown keff all rods inserted 0.96 0.96 Cold shutdown k ff rod of maximum 0.99 0.99 worth stuck fO out

1. 2. 3-3 TABLE 1.2.3:1 (Contd.)

PRINCIPAL FEATURES OF PLANT DESIGN Design Operating Standby liquid control shutdown, 0.16 dkeff Minimum Critical Power Ratio: 1.07 1.39 Linear Heat Generation Rate (kw/ft):

7x7 fuel 17.5 17.5 GE 8x8 fuel 13.4 13.4 ENC fuel 14.9 14.9 Hot Approximate Coefficients: Cold (no voids) Operating Moderator Tern~. Coefficient -8.9xl0- 5 -17.0xl0- 5

[ ( d k/ k ) I °FJ Moderator Void Coefficient less than_ 3 -1.0xlO -3 -1.4x10- 3

[ ( dk/k) /% Void]

Fuel Temp. (Doppler) Coefficient -l~~~~~l~ -1.2xl0- 5 -1.2x10- 5

[(dk/k)/°F]

Excursion Parameters:

1* Prompt Neutron Lifetime 48.9 microseconds

.B Effective Delayed Neutron Fraction .0058 Core Equivalent Core Dia. 182. 2 inches Circumscribed Core 189.7 inches Diameter Core Lattice Pitch 12 inches (4 assemblies/unit cell)

Number of Fue 1

,l\ssemb 1i es 724 Fuel Assembly GE GE GE ENC 7x7 8x8 8x8R/Px8x8R 8x8 Fuel Rod Array 7x7 8x8 8x8R/P8x8R P8x8 Fue 1 Rod Pitch 0.738 in. 0.640 0.640 0.641 Weight of U0 2 per 492.5 lbs. 458.6 441.6 434.4 Fuel Assembly Channel Material Zircaloy-4 Zircaloy-4 Zircaloy-4 Zircaloy-4 Total Assbly plus 678.9 lbs. 650 650 580 Channel Weight Fuel Rods 49 63 62 63 Water Rods 0 1 2 1

1. 2 .3-4 TABLE 1.2.3:1 (Contd.)

PRINCIPAL FEATURES OF PLANT DESIGN Fuel Rod, Cold GE GE GE ENC 7x7 8x8 8x8R/Px8x8R 8x8 Fuel Pellet Dia. 0.488 in. 0.416 0.410 0.405 Cladding Thickness 0.032 in. 0.034 0.034 0.035 Cladding O.D. 0.563 in. 0.493 0.483 0.484 Active Fuel Length 144 in. 144 145.24 145.24 Lgth of Gas Plenum 11.22 in. 11.24 9.48 10.06 Fuel Material U0 2 U0 2 U0 2 U0 2 Cladding Material Zircaloy-2 Zircaloy-2 Zircaloy-2 Zircaloy-2 Fi 11 Gas He He He He Fill Gas Pressure 1 atm 1 atm 1 atm/3 atm 3 atm Movable Control Rods Number 177 Shape Cruciform Pitch 12.0 in.

Stroke 144 in.

\4 i dth 9.75 in.

Control Length 143 in.

Control Material ~a c granules in stainless steel tubes and sheath Number of Cntrl Mtrl Tubes per Rod Tube Di mens i ans 0.188 in. o.d. x 0.025 in. wall Temporary Control Curtains Number 340 Shape Flat sheet Width 9.20 inches Thickness 0.0625 inches Control Length 141.25 inches Control Material Stainless steel containing 5400 ppm natural boron Curtain Locations Between fuel assemblies in water gaps without control rods.

Burnable Neutron Absorber Control Material Gd 2 03

• Location Mixed with U0 2 in several fuel rods per fuel assbly Concentration Location and reload dependent.

Reactor Vessel Inside Diameter 20 ft.-11 in.

Overall Length Inside 68 ft.-7-5/8in.

Design Pressure 1250 psig

Rev. 1 June 1983

• • • - TABLE OF CONTENTS SECTION 2 -- SITE 2i Page .

2.1 INTRODUCTION

2.1. 0-1

2.2 DESCRIPTION

Of SITE AND ADJACENT *AREAS 2. 2.1-1.

2;2;1 SITE 2.2.-1-1*.

2.. 2 .1.1 Site Size and Location , : .

2. 2.1-1 ...

• 2:. 2.1.2 Site Ownership . '*2 .. 2.1-1 2.2~1.3

• Location of the Units on the.Site 2.2.1-1:'

2.2.1.4' Oth~r Activities on. the Site . 2.2.1-2 2.2.1.5 Access to the Site

• 2.2.1-2 2.2.1.6 Exel us ion Area . 2.2.1-3;

2. 2. 2 . POPULATION AND LAND USAGE IN ADJACENT AREAS 2.2.2-1 2.2.2.1 Popu 1at ion Data 2:2.2.2 Land Use .. *

'2.2.*2.J POTENTIAL. HAZARn°S DUE TO)IEARBY FACilITIES .* ' *,* 2. 2~ 2::.6: *_<;, *: :. ':

2 2 '2 3 1 :INTRODUCTION: **.. * * .. . *.. * .* ' *. "* 2*; 2'. 2:;.6 .* '.**,*. *: :. ;*; ' ': <<

,. ' .. 2*:i:*2 . : f 2 *.. HAZARDS FROM* EXPLOSIONS. *. .* ..

• i . *. . 2.2.2..:.6'. *:

. ***e 2:2.2.3.2.1 *

• industrial Facilities*

.2.2.2.J .. 2.2 * ~ighway Transportati~n

2. 2'. 2. 3. 2. 3 Rail way Transportatfori 2~2.2-6 ;

.2.2*. 2-8 2... 2. 2~9 .

. 2.2.2~3.2.4 vJaterway Transportation 2. 2. 2;..10 2.2.2.3.2.S: Military Facilities 2.2.2-10.*.

2.2.2.3.2.6

• Pipelines 2. 2 .. 2-1 L

2. 2.. 2. 3. 3

• HAZARDS FROM. VAPOR CLOUDS AND FIRES 2.2. 2-n.,. *..

2.2.~2.3.4 *. HAZARDS FROM TOXIC CHEMICALS. 2 .. 2.2-11' 2.2.2.3.5 HAZARDS FROM COLLISION WITH THE INTAKE 2. 2.2-n STRUCTURE

. 2.2~.2.3.6 HAZARDS FROM .

LIQUID SPILLS. . .

z. 2. 2:.12 2*. 2 *. 2 *. 3. 7 HAZARDS FROM AIRCRAFT. 2.2.2-12

• *2. 2. 2. J. 7 .1 Airports 2.2.2~12 2.2 *.Z.3.T.2 Airways* 2. 2. 2-1.4 2 .. 2:.2*.

3.8 CONCLUSION

S 2.2.2;..15

• 2.2.2~

3.9 REFERENCES

2.2.2-16 I*.'.

Rev. 2 June 1984 2i ii LIST OF FIGURES -- SECTION 2, SITE 2.2.1:1 Station Property Plan 2.2.1:2 Cooling Lake General Arrangement 2.2.2.3:1 Dresden Nuclear Power Station Area Map 2.2.2.3:2 Pipelines Considered in the Evaluation of Hazard From Explosion 2.2.4:1 Cooling Water Flow Diagram -- Unit 2/3 2.2.4:2 Dresden Cooling Lake Dam 2.2.6:1 Site Flow Diagram at Illinois River Above Dresden Island Lock and Dam 2.2.6:2 General Arrangement .-- Crib House

2iii LIST OF TABLES -- SECTION 2, SITE 2.2.2:1 Population Centers .Surrounding Station 2.2.2:2 Industrial Facilities Near Station 2.2.2:3 Recreational and Institutional Facilities Near Station 2.2.5:1 Distances From Release Points To Various Points Near Site

.I.:

• • e* .......

• • Table 2.?.*2~_3:l Assessment Summary

.' '1 HAZARD .*. REPORT .

• NUMBER SOURCE OF POT~NTIAL HAZARD *... SECTION*. DESIGN BASIS EVENT?

Explosion from:*.

l* Industrial facilities* ,2.1 . No, based Qh adequatE? separation distance**

2 Highway transportation *2 .. 2 ' No ? based on adequ~te separation distance 3 Railway transportation *2:3' ' No, .based on. adequate separation distance 4 Watt;!rway transportation '2.4' No, based on adequate separation di~tance 5 Military facilities ,* 2. 5 No, based .on adequate separatigg distance 6 Pipelines 2.6 No, based on frequency of 6x10 /yr using conservative ~ssumptions Vapor cloud expiqsion & fire from waterway transporation 3.t No, pased on freq~en~Y of 4xl0 -7 /yr 8 Toxic chemicals 4 Not part of SEP U-1.C 9 Collis{on with intake structure 5 Nq, based on physii:al considerations 10 Li quid spi 11 s .*~. *.No, based on physical considerations Aircraft .impact from:

'.< . *; ~*. ' .** *.. ; -7 '

11 Airports 7:*1: No, based on frequency of 3.24x10 l¥ear 12 Airways* T:f .~. No~ based on .f~egu~ncy of 0.93 x 10 /year

'l'rData for facilities which responded to. the q~estion'nalre," * .

• • There is one exception to this conclusion .:. ~he:benie~¢ .storage tank pn the Reichhold Chemical site.*  :* * .. ; * <r .

-~* ...

• , e. Table 2.2.2.3:2 Industries Within 5 Miles Dresden Station (Ref. 18)

I DISTANCE (MILES)

INDUSTRY & DIRECTION . PRODUCT GE BWR Training Center

&Spent Fuel Storage 0. 7 -: SL~ Spent nuclear fuel storage Reichhold Chemicals 1. 6. - w Resins and chemicals A.. P. Green-* .

• 2. 1 "". SSW-* Br.iCk and clay Atrco 1ndustrial Gases

• 2.S NW co 2 . .

I.

Northern Illinois Gas

• 2*,5_,.. NW Natl,J na 1 gas Alumax Mill Products 2.8 - MW Aluminum sheet and co.il

• Northern Petrochemicals 3. 3 -* MW Ethy\ene~ ethyl en~ oxide glycol*

Northern Petrochemical Dock . 2~ 1 -* W*

~*. .

• ARMAK Chemicals 3.6 - WNW Fatty nitrogen chemi.CaJs

• Dur.kee scM~ Chemicals  :. *.*, J~ 2>- .EN{ .; Ed.i b le. oi-l

, Truck Tennina*i J.6*;,. ENE

• Under construction *.

• Dow Chemicals . 3. 7 - E .. Polystyrene** pla-stic.

• Dow Chemical *Dock . 2. 7 -* E

• • ~ *.

'Exxo_n (chemical plant) J.9 - ENE Under construction Hydrocarbon Transportation, Inc. 4.0 ..: NW .Propane Streator Industrial Supply 4. 0 - .s Industrial supplies Mobil Chemical Co~ -4.1 - NE . Po.lystyrene sheets. & crystal Jal iet Livestock Market 4.2 - ESE Livestock

• Mo_bn O:il Refi-nery .. 4. 5 - NE Petroleum** products Commonweal th Edison Co *

. Collins St~tion 5 *. 0 - WSW Electricity*

~  :.::..; *.:'".{ . .\*... . .., ._..... ** ....... .

e *.*

. , a.,, .* t *** '.:.:* * . * . ; .. . *' l'.:. :-.:~ .. i ..*. ; ' ' ' ' .. .  :~ - . '

Table2.2.2.3:3 Dresden Island Traffic Statistics Fiscal Years 1~73:~ l9j8 (Refs. 6, 11)

CO~MODITY,TYPE . FISCAL YEAR 1973 1974 1975 1976 1977 1978 Average .

Total commodities, tons x 10 28.476 30.853 27.808 25.882 23.452 19. 521 . 26.0 Hazardous mate5ials,* ~-

• tons x 10 . 5.653 . 6.073 5.358 5.059 4.093 3.658 5.0 Liquefied Gases,** tons o.o* 0.0 O*.O *

• 17 ,992 0.0 0.0 . 3000.0

• Hazardous materials are defined as all materials listed under the.

category of petroleum products in the lock statistics.

• • Liquefied gases shown are the amounts transported on the entire navigable length of *the lliinois River.

Table 2.2.2.3:4 Casualty and Spill Statistics -

Fiscal Years 1969 thru 1972 (Ref. 10)

ILLINOIS WESTERN CASUALTY/SPILL RIVERS *RIVERS Casualties** - all type barges 178 2831 Casualties of hazardous material barges***. 40 508 Spills from hazardous

• .mat.erial barges 1 69 Casual ti es* of Liquefied gas barges ._.;._

9 Spills from double-skinned vessels 7

... Total length of waterway (miles) 333 3137

• Lower Mississippi, Upper Mi-ssiss.ippi, Ohio, and Illinois Rivers; casualties from these rivers* constitut~ 97% of the casualties on western rfvers. .,

• • Casualtie.s whfch result in any of the following: loss of life,.

damage to cargo-irr excess of $1;500, or release of cargo. ~

• • • Hazardous material barges are generic type 17, 18, and 29 vessels.

See Reference 10 for description.

TABLE '2.2.2.3:5 DATA ON AIRPORTS WITHIN 10 MILES OF DRESDEN STATION.(REFS .. 22~ 23, 27)

APPROX. DIST. DIRECTION NO. LENGTH OF WIDTH OF TYPE ORIENTATION TYPE FROM STATION FROM STAT ION OPERATIONS. RUNWAY RUNWAY OF RUNWAY OF RUNWAY FROMM PVT. 4.5miles E 50* 2,773 ft. 100 ft. TURF. NNE-SSW MORRIS PVT. 8 mil~s WNW 1~94.2* 2,400 ft. 135 ft. TURF. E-W

?,987 ft. 60 ft. ASPH. N-S ROSSI PVT. 9 miles N 50** 2,400 ft. 70 ft. TURF. E-W BUSHBY PVT. 9.9 miles NNE 45** 1,800 ft. 100 ft. TURF . N-S

.JOLIET . Pub. 10 miles NNE 10,000* 3,452 ft. 125 ft.* TURF. NE-SW 2~970 ft. 100 ft. ASPH. NW-SE ADELMANN*** PVT. 1 mile NE 20*'11' 1,600 ft. 70 ft. TURF. SE-NW

• Total peak month from FAA supplied documents.

• • Number per month as supplied by owner of airport

• • • Recent1y approved airstrip

e e Table 2.2.2.3:6 Data for Aircraft Crash and Probability Analysis 6 N OPERATING r 0 D(r,O) 2 .x 10 R (OPERATIONS/ A NARDx1C'i 7

. AIRPORT MODE (MILES) (DEG) (/MILES ) (/OPERATION) YEAR) (MILES 2) (/YEAR)

FROMM Landing 4.5 90 0.0014 2~4 150 0.0056167 0.02833 4.5 90 0.0014 2.4 150 q.0056167 0.02833 Take-Off 4.5 90 0.00167 0.9 150 0.0056167 0.01267 4.5 90 0.00167 .a~ 9 150 0.0056167 0.01267 MORRIS Lanqing 8.0 25 . 0. 000883 :. . 2.4 1456 0.0056167 0.17333 8.0. 155 0.000043 2A 1456 0.0056167 0.00833 8.0 65 0.00035 2.4 4370 0.0056167 0.206 8.0 115 o. 00011 2.4 4370 0.0056167 0.06467 Take-Off 8.0 25 . 0.000369 0.9

..~ ' 1456 . 0.0056167 0.027 8.o 155 0.000073 0.9 1456 0.0056167 0.00533 8.0 65 *._ o. 00022 0. 9* 4370 0.0056167' 0.04867 8.0 115 0.00012 0.9 4370 0.0056167 0.02633 JOLIET Landing 10.0 . 10 0.00045 , *2A .* 6000. 0.0056167 0.364 10.0 170 0. 000011 . *. 2. 4 9000 . 0. 0056167 0.01333 10.0 80 0.000088 2.4 22500 0.005p167 0.26667 10.0 100 o. 000056 . 2.4 22500 0.0056167 0.17 Take-Off 10.0  : 10 0.00013 . 0.9 *. 7500 0.0056167 0.04933 10.0 170 0.000018 b. 9' 7500 . 0.0056167 0.00667 10.0 80 0.000055 o.~ 22500 0.0056167 0.06267 10.0. 100 0.000043 0.9 22500 0.0056167 0.049 ADE~MANN Landing 1.0 115 0.01433 2.4 60 0.0056167 0.116 0.9 80 0.0374 . 2. 4 60 0.0056167 0.30233 Take-Off LO 115 0.0317 0.9 60 0.0056167 0.0906 0.9 80 0. 05734 . 0 *. 9 60 0.0056167 0.174

,*.JO.

Table 2.2.2.3:7 Pi~elines with 5. Miles of the Site PIPE SIZE OPERATING CLOSEST DISTANCE PIPELINE COMPANY (in) MATERIAL CARRIED PRESSURE (PSI) TO THE PLANT (MILtS)

Natural Gas 36 Natura 1 Gas 858 \ 1. 75 Pipeline Co. 36 **Natural Gas 858 1. 70 30 Natura 1 Gas 858 1.60 36 Natural Gas 650 1. 25 30 Natura 1 Gas 858 1. 70.

30 Natural Gas 858 1.60 Hydrocarbon 10 Propane, Natural. Gas 2100 4.0 Transportation, Inc. 10 Propane, Natural Gas 2100 4.0 6 Propane, Butane 500 2.0 Northern lllinois Gas 36 Natural Gas 740 2.5 10 Out of Operation 2~5 4 Natural Gas Unknown 3.0 Amoco 10 Crude Oi 1 3.0 12 Crude Oi 1 3.0 22 Crude Oi 1 3.0

• .'l-_

• 1 ,. :

e *,,'

INDUSTRIAL SITES IN VICINlfY 1 MlOWEST FUELS REPROCESSING PLANT {GE).

2 NORTllER~ PETROCHEMICAL CO.

3 /\LUMAX 4 REICllHOLO CHEMICAL CO

5. A. P. GREEN*

6 GENERAL ELECTRIC CO.TRAINING SCHOOL 1 MOBIL CHEMICAL 8

• MOBIL OIL 9 DURKEE SCH

saL\ET ~l?.~Y t\MMUN ITION ~L.1'NT t I *I

... PeN .1 Mill~

FIGURE 2.2.2.3:1 DRESOEN NUCLEAR POWER STATION AREA MAP

. -~ .. *.

. . =**

• . - . ~ *' '.**. . * ....

• i*'

, -*~ *.

LEGEND "*. .

36" radius

-.~.- 36"

.. -~~- JO"

- - 36"'

' ...... 30 11

• JO"

., *SITB

"" . "'.**0*':**

• .. **: ~. :. .

.1.*

'.  : .:- . *' ~

FIGURE 2.2.2.3:2 PIPELINES

. . CONSIDERED

. / . . ' IN

'. .... :rnE .. '.. EVALIJAnoN .. ' . '

OF

. HAZARO FROM EXPLOSION'

. ...